JPS5815079B2 - Radioactive waste disposal method from nuclear fuel reprocessing facilities - Google Patents

Description

[Detailed Description of the Invention] The present invention provides a nuclear fuel reprocessing facility (hereinafter referred to as a reprocessing facility)
The present invention relates to a method for reducing the volume of soluble and insoluble radioactive waste generated from.

Nuclear fuel reprocessing involves separating and recovering the uranium and plutonium contained in the nuclear fuel used in a nuclear reactor and processing it so that it can be used again as nuclear fuel. Currently, Purex is a wet method. It is well known that the method is adopted in most reprocessing facilities.

FIG. 1 is a block diagram showing the steps of a reprocessing facility using the Purex method and the radioactive waste generated from each step.

In the Purex method, spent nuclear fuel is first subjected to pretreatment process 1.
In step 5, the fuel solution is cut into several lengths, firstly it is dissolved in nitric acid in a dissolution step 16, and then in a separation and extraction step 17, this fuel solution is converted into tributyl phosphate (TBP) using a mixer settler or a pulse column. The fission products (FP), uranium, and plutonium are separated by contact with the flow, followed by extraction, washing, and back-extraction operations, and then in the purification process (uranium purification process 18 and plutonium purification process 19), uranium and plutonium undergo ion exchange. Alternatively, it is purified by adsorption operation.

In this way, the Purex method uses chemical treatment, so
Each process generates some radioactive waste.

First, regarding solid waste, nuclear fuel cladding material that does not dissolve in nitric acid generated from the melting process 16, used ion exchange resin and adsorbent generated from the refining process, and miscellaneous items such as paper and clothing used within the facility. There are solids.

Among these solid wastes, the coating material has a high radioactivity of about 400 Kyu per ton of fuel, so it is stored as is in storage as high-level solid waste20.
However, combustible miscellaneous solids 22 are incinerated in an incinerator 23, and non-combustible solids such as ash from the incinerator, used ion exchange resins and adsorbents are compressed and reduced in volume.
It undergoes concrete solidification24 and other treatments and is stored in low-level solid waste storage.

Next, regarding liquid waste, there are high and medium level waste liquids 25 and 26 generated from the extraction and separation process 17, and low level waste liquid 27 such as washing water and miscellaneous liquid generated from within the facility.

The amount of high-level waste liquid 25 generated is approximately 10 m per ton of fuel.
3, which contains as high as 106 Ci of radioactivity.

This is because this waste liquid contains 99.9 or more FP in the nuclear fuel.

The high-level waste liquid 25 has high radioactivity and solidification methods have not been put into practical use, so it is evaporated and concentrated using an evaporator.
Afterwards, it is stored in a high-level waste liquid storage tank.

Solvents such as TBP used in the extraction and separation step 17 are damaged by radiation and their performance deteriorates, so they are washed with an alkali such as caustic soda and reused.

The cleaning solution, caustic soda, is included in the medium-level effluent along with nitric acid.

Sodium nitrate is produced by the reaction between caustic soda and nitric acid, so the main component of medium-level waste liquid is sodium nitrate.

The amount of medium-level waste liquid generated is about 50 m3 per ton of fuel, which contains about 0.1 mol/l of sodium nitrate and about 102 Ci of radioactivity.

The medium-level waste liquid 26 is treated by evaporation 30 to remove moisture and coagulation and precipitation 29 to remove radioactive substances.

The sodium nitrate concentrate whose volume has been reduced by the evaporation operation is solidified 31 by asphalt solidification or cement solidification, and is stored in a storage.

Further, the sludge generated in the coagulation and sedimentation 29 is dehydrated and reduced in volume in a filtration device, and then similarly solidified 32 and stored.

Regarding the low-level waste liquid 27, the content of sodium nitrate and radioactivity is smaller than that of the medium-level waste liquid 26, but since the amount generated is large at approximately 150 m' per ton of fuel, it cannot be treated in the same way as the medium-level waste liquid. It is being done.

As mentioned above, at reprocessing facilities, various types of radioactive waste are treated differently, including incineration, compression, volume reduction, evaporation, coagulation and precipitation, and solidification, depending on the radioactivity level and properties. There is.

Therefore, a wide variety of equipment is required, which increases the equipment capacity and complicates operation.

Operating and maintenance costs also increase significantly.

Furthermore, since the form of the product differs depending on the processing method, various storage facilities are required.

In addition, since medium- and low-level waste liquids contain nitric acid, there is a problem in that evaporators and solidification equipment are susceptible to corrosion.

An object of the present invention is to collectively process radioactive waste generated from a reprocessing facility to simplify the processing equipment.

A feature of the present invention is that various radioactive wastes generated from reprocessing facilities are collectively dried and powdered, and then molded to reduce the volume.

A preferred embodiment of the present invention will be explained in detail with reference to FIG.

FIG. 2 is a flowchart of the method for treating radioactive waste from a reprocessing facility according to the present invention.

Here, in this reprocessing facility, spent fuel from a light water reactor is processed using the Purex method at a rate of 1 ton/day.

In the figure, 1 is a waste liquid receiving tank, 2 is a sludge receiving tank, 3 is a waste liquid resin receiving tank, 4 is an incinerated ash receiving tank, 5 is a processing tank, 6 is a slurry pump, 7 is a centrifugal thin film dryer, 8 is a motor, and 9 is a scraper. feather,
10 is a heating steam jacket, 11 is a granulator, 12 is an asphalt receiving tank, 13 is an asphalt supply pump, and 14 is a drum.

In this embodiment, a centrifugal thin film dryer 7 is used as the dryer, which has a cylindrical heat transfer surface inside and rotary blades 9 rotated by a motor 8 so as to be in contact with the heat transfer surface. The heat transfer surface, which is a part of the body of the centrifugal thin film dryer 7, is heated by passing steam through the outer heating steam jacket 10.

The lower part of the centrifugal thin film dryer 7 is connected to the upper part of the granulator 11.

The lower part of the granulator 11 is connected to the upper part of the drum 14.

In addition, an asphalt receiving tank 12 is provided on the upper part of the drum 14.
Asphalt can be supplied from the lower part of the tank via an asphalt supply pump 13.

Radioactive waste generated from each process at the reprocessing facility will be treated as follows.

Medium and low level waste liquid 26 generated from the extraction and separation process 17
and 27 are collected in the waste liquid receiving tank 1.

If this waste liquid contains volatile substances such as ruthenium (Ru), the radioactivity decontamination coefficient in the centrifugal thin film dryer 7 will decrease.

Therefore, the concentration of volatile substances such as Ru is measured using a radioactivity analyzer (not shown), and if this value exceeds a standard value, the waste liquid is transferred from the waste liquid receiving tank 1 to the sludge receiving tank 2.

Then, a coagulating precipitant such as lime soda is added to this waste liquid to remove volatile substances.

Medium- and low-level waste liquids contain nitric acid and caustic soda, but the former is usually more abundant, so the waste liquid is acidic with nitric acid.

Metal materials have a tendency to corrode due to the acidity of nitric acid.

In order to prevent starvation, it is possible to add a neutralizing agent to this waste liquid, but this increases the amount of waste.

Plutonium purification processes usually use cation exchange resins, which have the property of neutralizing the acidity of nitric acid.

The main component of incineration ash is calcium oxide, which has the property of becoming alkaline when dissolved in water.

Therefore, the low-level waste liquid in the waste liquid receiving tank 1, the waste resin in the sludge waste resin receiving tank 3 of the sludge receiving tank 2, and the incinerated ash in the incinerated ash receiving tank 4 are mixed in the processing tank 5.

The pH of the liquid in the treatment tank 5 is measured with a pH detector, and the amounts of waste resin and incinerated ash supplied are adjusted so that this value falls within a predetermined range.

The treatment tank 5 that has been neutralized in this way becomes slurry at night.

This slurry mixture is supplied to a centrifugal thin film dryer 7 by a slurry pump 6.

The slurry mixture supplied to the centrifugal thin film dryer 7 is subjected to centrifugal force by the rotary scraping blades 9, and is discharged as powder by the rotary scraping blades 9.

Since the centrifugal thin film dryer 9 has a strong centrifugal force, good powder can be obtained even from slurry mixed with waste resin, sludge, and incineration ash.

The powder is processed into pellets by a granulator 11.

Since the ion exchange resin and incineration ash act as a binder for the pellets, the pellet strength of the slurry mixture is greater than that of sodium nitrate alone, and the addition of these components improves the properties of the pellets.

The processed pellets are discharged into a drum 14.

Asphalt receiving tank 1 in drum can 14 filled with pellets
When asphalt is poured from 2 through the asphalt supply pump 13, solidified asphalt pellets are obtained.

This solidified material is stored and stored in a storage.

Since the specific gravity of this solidified material is 1.2 or more, it can be adequately used for future ocean dumping.

As described above, according to this embodiment, the low-level waste liquid, waste resin, and incineration ash generated from the reprocessing facility can be volume-reduced all at once, which simplifies the processing facility.

Further, since the neutralization treatment of waste liquid can be performed without increasing the amount of waste, material waste in the centrifugal thin film dryer is prevented, and the load on the centrifugal thin film dryer is reduced.

As another embodiment of the present invention, even if the pellets are solidified with plastic, they have the same effect as solidified asphalt.

As described above, according to the present invention, the following effects can be achieved.

(2) Since various types of radioactive waste generated from reprocessing facilities can be treated all at once, the processing equipment is simple and operation control is easy.

2. The life of the centrifugal thin film dryer is extended and the load is reduced.

3. Storage equipment is standardized because the final form of waste is pellets with the same shape.

[Brief explanation of the drawing]

Figure 1 is a block diagram showing a method for treating radioactive waste generated from reprocessing facilities using the conventional Purex method;
The figure is a radioactive waste treatment system diagram that is an embodiment of the present invention. 5... Processing tank, 7... Centrifugal thin film dryer, 11... Granulator.

Claims (1)

[Claims] 1 Mix nitric acidic waste liquid generated from nuclear fuel reprocessing facilities, waste resin that has the effect of neutralizing the waste liquid, and incineration ash of combustible waste so that the mixed liquid becomes neutral or almost neutral. A method for treating radioactive waste from a nuclear fuel reprocessing facility, which comprises adjusting the mixed liquid into a dry powder using a centrifugal thin film dryer, and reducing the volume by forming the powder into pellets.